Office of Compensation Analysis and Support 

Technical Basis Document for Atomic Energy Operations at 

Blockson Chemical Company, Joliet, Illinois 

Subject Experts: Tom Tomes, Sam Glover 

9/11/2006 Approval: Signature on File 

Date: 

James W. Neton, Associate Director for Science 

Document Number: 

OCAS-TKBS-0002 

Effective Date: 9/11/2006 

Revision No.: 0 

Controlled Copy 

No.:________ 

Page 1 of 27 

Supersedes: 

ORAUT-TKBS-0002 

TABLE OF CONTENTS 

Section Page 

1.0 Introduction..........................................................................................................................3 

2.0 Site History ..........................................................................................................................4 

3.0 Internal Dose ........................................................................................................................7 

3.1 Uranium Intakes.......................................................................................................7 

3.2 Radon Exposures ...................................................................................................11 

4.0 External Dose .....................................................................................................................11 

4.1 Source Term ...........................................................................................................12 

4.2 Exposure from Drums of Uranium ........................................................................13 

4.3 Exposure from Contaminated Surfaces .................................................................17 

4.4 Beta Dose ...............................................................................................................18
 

4.5 Occupational Medical Dose ...................................................................................19 

5.0 Dose from Residual Contamination...................................................................................20 

6.0 Dose Reconstruction Summary..........................................................................................24 

7.0 References ..........................................................................................................................25 
 

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LIST OF TABLES 

Table Page
 

1a Inhalation rate for operations .............................................................................................10 

1b Ingestion rate for operations ..............................................................................................11 

2 Radon exposures ................................................................................................................11 

3 Activity of aged uranium following separation and 100 day ingrowth .............................14 

4 Uranium dose rates from drums of yellowcake .................................................................14 

5 Annual organ dose from exposure to drums of yellowcake ..............................................17 

6 Shallow dose ......................................................................................................................19 

7 Annual organ dose from medical X-rays ...........................................................................19 

8 Annual dose from residual contamination .........................................................................20 

9a Inhalation intake from residual contamination ..................................................................22 

9b Ingestion intake from residual contamination....................................................................23 

LIST OF FIGURES 

Figure Page 

1 Photograph of the Joliet plant of the Blockson Chemical Company .................................. 5 

2 The Blockson Chemical process for the recovery of uranium from phosphoric acid .........6 

3 Distribution of uranium intake rates ..................................................................................10 

4 Beta dose rate on the surface of yellowcake ......................................................................13 

5 Beta dose rate from yellowcake separated from ore for more than 100 days as a 

function of distance from the surface.................................................................................13 

6 Effect of density on dose rate at 30 cm from drum of U3O8..............................................15 

7 Cumulative dose rate from photons and Bremsstrahlung at 30 cm from a drum 

containing U3O8 (density 4 g/cm3) ....................................................................................16 

RECORD OF ISSUE/REVISIONS 

ISSUE 

AUTHORIZATION 

DATE 

EFFECTIVE 

DATE 

REV. 

NO. DESCRIPTION 

Draft 6/29/2006 00-A Document initiated to establish the 

technical basis for radiation dose 

reconstruction for former workers from the 

Blockson Chemical Company. 

7/31/2006 9/11/2006 00 Incorporates resolution of review 

comments. This supersedes ORAUTTKBS-0002. A PER is required for dose 

reconstructions performed with the 

superseded document. 
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1.0 Introduction 

Technical basis documents and site profile documents are not official determinations made by 

the National Institute for Occupational Safety and Health (NIOSH) but are rather general 

working documents that provide historic background information and guidance to assist in the 

preparation of dose reconstructions at particular sites or categories of sites. They will be revised 

in the event additional relevant information is obtained about the affected site(s). These 

documents may be used to assist NIOSH staff in the completion of the individual work required 

for each dose reconstruction. 

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic 

weapons employer facility” (AWE facility) or a “Department of Energy facility” as defined in 

the Energy Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 

U.S.C. § 7384I(5) and (12)]. EEOICPA, as amended, provides for employees who worked at an 

AWE facility during the contract period and/or during the residual contamination period. 

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when 

the AWE was processing or producing material that emitted radiation and was used in the 

production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination outside 

of the period in which weapons-related production occurred). For contract period employment, 

all occupationally-derived radiation exposures at covered facilities must be included in dose 

reconstructions. NIOSH does not consider the following exposures to be occupationally-derived: 

• radiation from naturally occurring radon present in conventional structures; and 

• radiation from diagnostic X-rays received in the treatment of work-related injuries. 

For residual contamination period employment, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) (i.e., radiation doses received from DOE/AEC-related work) must be 

included in dose reconstructions. Radiation dose received from DOE/AEC-related work 

includes: (1) radiation from radon consistent with NIOSH’s policies for including such radiation 

in the contract period; and, (2) medical screening X-rays, but not diagnostic X-rays for the 

treatment of work-related injuries. It should be noted that: (1) under subparagraph A of 

§ 7384n(c)(4), radiation associated with the Naval Propulsion Program is specifically excluded 

from the employee’s radiation dose; and, (2) under subparagraph B of this section, radiation from 

a source not covered by subparagraph A that cannot be reliably distinguished from radiation that 

is covered by subparagraph A is considered part of the employee’s radiation dose. This site 

profile covers only exposures resulting from nuclear weapons-related work. Exposures resulting 

from non-weapons related work, if applicable, will be covered elsewhere. 

Section 2.0 of this document describes the Blockson Chemical Company site and its history 

including some information about the radiological processes and source terms as well as the 

radiological controls and monitoring practices. Sections 3.0 and 4.0 discuss internal and external 

dose, respectively. 
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The following summary is to help provide consistency in dose reconstructions and to help assure 

that all components of dose are adequately addressed when doses are reconstructed. It also 

provides some information regarding the radiological processes and source terms, as well as 

information on the radiological controls and monitoring practices. While not meant to substitute 

for a complete site profile, it represents the best understanding of the covered site at this time and 

provides assumptions for estimating doses when specific dose-related information is not 

available in individual records. 

This site profile provides specific information on historical practices and radiation exposures at 

the Blockson Chemical Company facility in Joliet, Illinois. Blockson Building 55 and related 

activities are covered under EEOICPA (DOE 2006). 

2.0 Site History 

In 1950-1951 the U. S. Atomic Energy Commission (AEC) approached several phosphate rock 

consumers about the possibility of recovering the uranium from the phosphate rock they 

processed. At the Blockson Chemical Company plant, the AEC was interested in the uranium 

that could be separated from the phosphoric acid. In early 1951 the research staff at Blockson 

began an evaluation of the available research data and preliminary experimentation that the AEC 

made available to them. They determined that the only economically feasible approach 

applicable to the Blockson process would be to make the uranium recovery a by-product process 

(Stolz, Jr. 1958). 

On March 6, 1951, the AEC entered into letter contract number AT(49-1)-606 with Blockson 

Chemical Company to develop a process to extract uranium from wet phosphoric acid (DOE 

1983, DOE 2006). This letter contract was later replaced by contract number AT(49-1)-611 on 

October 18, 1951 (AEC 1951). Under the contract, Blockson constructed Building 55, at its own 

expense, to house uranium recovery equipment at their plant in Joliet, Illinois. The uranium 

recovery plant was constructed to recover uranium from phosphoric acid being produced by 

Blockson from normal commercial operations. On August 15, 1952, Blockson began production 

and delivery of uranium concentrates to the AEC (Stolz, Jr. 1958). 

In 1955, Blockson was sold to the Olin Mathieson Chemical Corporation who assumed the 

liabilities and obligations of Blockson under all contracts, as stated in contract number AT(49-1)­

611 Amendment 1. The Olin Corporation continued the uranium recovery program under 

contract with the AEC. In March 1962, the uranium recovery work was discontinued with the 

expiration of the contract (DOE 1983). 

According to the contract signed in October of 1951, Blockson, and later Olin Mathieson, was 

responsible for the health and safety of the employees at the site and for conforming to AEC 

health and safety regulations and requirements. In Amendment 3, effective January 1, 1958, this 

statement was deleted. Personnel with the Formerly Utilized Sites Remedial Action Program 

(FUSRAP) conducted records searches for information regarding the uranium recovery activities 

at Blockson. No records of health and safety inspections by the AEC were found as a result of 

their search, although there was evidence of periodic visits by AEC personnel to review and 

audit process operations (DOE 1985). Records of bioassay monitoring have been found during 
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data capture efforts for the EEOICPA program, but there is no indication from records searches 

that external dosimetry was utilized at Blockson. 

The recovery plant was put into operation on August 15, 1952. The process was patented and 

the patent, USP 2743156, was assigned to the AEC (Stolz, Jr. 1958). Building 55 was a onestory, 100-by-175-foot building built specifically to house the uranium recovery process (AEC 

1951, DOE 1983). A photograph of Building 55 is show in Figure 1. 

Figure 1: Photograph of the Joliet plant of the Blockson Chemical Company. 

[Reproduced from Barr et al. 1955]. 

The Blockson Chemical Company manufactured wet-process phosphoric acid from Florida 

phosphate rock (Barr et al. 1955, Clegg and Foley 1958). Figure 2 shows a schematic flowchart 

of the Blockson process for the recovery of uranium from wet phosphoric acid. 

The plant produced technical phosphates rather than fertilizers from wet phosphoric acid 

(Wilkinson 1976). In the Blockson process, the phosphate rock was calcined and then digested 

with sulfuric acid resulting in phosphogypsum and phosphoric acid. The phosphogypsum 

partitions most of the calcium and radium, and the phosphoric acid partitioned about 90% of the 

uranium. Very little uranium was lost to the phosphogypsum. The phosphoric acid was then 

converted into monosodium phosphate and other phosphorus derivatives. The uranium byproduct was precipitated from the monosodium phosphate stream. The monosodium phosphate 

liquor was heated and clarified. Sodium hydrosulfite (Na2S2O4) was added to precipitate the 

uranium. The liquor was filtered and the filtrate returned to the phosphate-processing plant. The 

precipitate, containing about 5% U3O8 was slurried in water in an upgrading step in which the 

uranium was redissolved. The uranium was then reprecipitated as uranous phosphate. The 

slurry was filtered and the precipitate, known as yellowcake and containing 40 to 60% U3O8, was 

dried for shipping (Clegg and Foley 1958, McGinley 2002, Wimpfen 2002). The uranium 

content of the phosphate rock consumed in these processes averaged about 0.014% U3O8 (Stolz, 

Jr. 1958). 
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Figure 2 The Blockson Chemical process for the recovery of uranium from phosphoric 

acid. [reproduced from Clegg and Foley 1958]. 

A pre-operational letter from Blockson to AEC in July 1951 provided estimates of construction 

and operating costs based on Blockson processing 6,000 tons of phosphate rock per week 

(Blockson 1951). Taking recovery into consideration they estimated producing 50,000 pounds 

of U3O8 per year. AEC reported in 1955 that U3O8 production from Blockson was 3,758 pounds 

in August; 3,407 pounds in September; 5,908 pounds in October; 4,093 pounds in November; 

and 2,937 pounds in December of 1955 (AEC 1955a, 1955b, 1955c, 1955d, 1955e). This 

indicates an average production rate of over 4,000 pounds per month in the latter part of 1955. 

The December report included information that Blockson had produced a total of 121,400 

pounds in the 40 months of Blockson operations to that point, an average of about 3,035 pounds 

of U3O8 produced per month through 1955. The total amount of uranium produced from 1955 

through 1962 is not precisely known. Per a contract amendment in 1958 production was limited 

to 50,000 pounds of U3O8 per year (US DOE 1985). 
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3.0 Internal Dose 

Workers in Building 55 and related activities were potentially exposed to airborne uranium. 

Thorium, assumed to follow the uranium in the Blockson process, and radon are also considered 

to be present in Building 55 and is included in this document. 

3.1 Uranium Intakes 

The greatest potential for internal uranium exposure in the Blockson uranium recovery process is 

associated with handling dried uranium compounds in the packaging areas. Here the uranium 

concentrate (yellowcake) was dried and barreled for shipping, resulting in a potentially dusty 

operation (Eidson and Damon 1984, US NRC 2002a, Wimpfen 2002). In all other areas of the 

Blockson plant, wet processes were used, and the surface contamination and potential dust 

exposures are minimal in comparison with the drying and packaging operations (Clegg and 

Foley 1958, US NRC 2002b). 

There are no air sampling results available from Blockson to characterize airborne concentration 

of uranium, however, air samples results from mills with some common activities have been 

published. A study was done (Eidson and Damon 1984) of uranium aerosols generated during 

yellowcake packaging operations at four uranium mills. The precipitated yellowcake is 

dewatered in a filter process, and then dried in an oven to produce powdered yellowcake that is 

placed in 55-gallon drums. Blockson also used a dryer to dewater the yellowcake prior to 

packaging (Clegg and Foley 1958, Blockson 1951). 

Eidson and Damon’s study described a sequence of steps common to all four uranium ore 

processing mills: 

1.	 No activity. This is when the plant is shut down for maintenance or when all available 

yellowcake has been barreled. Worker exposure to airborne yellowcake is minimal at 

this step. 

2.	 Barrel loading. This occurs when a barrel is placed under a hopper containing the dried 

yellowcake. The yellowcake is allowed to fall into the barrel. The amount of time 

workers spend in this area depends on the volume of the yellowcake in the hopper. (It is 

unclear whether a hopper was used at the Blockson Chemical facility.) 

3.	 Barrel uncovering. This step occurs when a filled barrel is removed from beneath the 

hopper. In some cases, the barrel may be vibrated to compact the yellowcake before 

removing the barrel from beneath the hopper. (It is not known of the barrels at Blockson 

were vibrated.) 

4.	 Powder sampling. This occurs when a worker takes a sample of yellowcake to analyze 

for moisture content. 

5.	 Lid sealing. This occurs when a worker places a lid on the barrel and seals it. 

6.	 Other activities. This step includes maintenance and cleaning of the area with water 

hoses. 
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During the study, air samples were taken in yellowcake packaging areas before, during, and after 

barrels of yellowcake were filled and sealed. Median air concentrations during the study ranged 

from 0.04 to 0.34 μg U/L (40 to 340 μg U/m3). 

Comparison of the above uranium mill results to Blockson operations cannot readily be made 

due to differences in operations and in quantities produced. Blockson produced a small quantity 

of yellowcake from phosphoric acid. Based on the reported masses produced, just a few drums, 

possibly up to 5 per month, would have been produced, assuming a nominal 1,000 pounds of 

yellowcake per drum. The mills processed uranium bearing ores that contained larger relative 

uranium concentrations, and in typically larger quantities. 

Although there are no air sampling results available for estimating intakes to Blockson workers, 

urinalysis results for uranium are available for Blockson workers, which is the preferred method 

of estimating internal dose received from exposure to uranium. Bioassay results are not 

available for individual claims submitted to NIOSH for dose reconstruction under EEOICPA, 

although the names on available results match a few workers whose job at Blockson is known. 

Default inhalation (or ingestion) intakes are presented below to be used for dose reconstructions 

when bioassay results are not available for a claim. 

The bioassay results were evaluated and assessed to provide intake rates that are favorable to 

claimants and are considered bounding for Blockson workers. This evaluation assumes two 

categories of workers. Workers that are considered to have the highest potential for exposure are 

categorized as production workers in this evaluation. This category includes operators who 

handle and package dried yellowcake daily. Since there is no definitive data to differentiate 

exposure rates to production workers who are exposed to the highest concentrations and those 

exposed only intermittently, e.g., maintenance mechanics, all production workers are assumed to 

be exposed to the bounding concentration. The other category of exposure in this evaluation are 

administrative workers. The intake rates for administrative workers are based on the assumption 

that they would not be exposed to any significant degree in close proximity to the dried 

uncontained material, but could have been exposed to elevated levels of general area airborne 

uranium contamination on a continual basis. 

Records indicate that Blockson employed approximately 20 people in the uranium recovery 

operations. In 1951, prior to start up of operations, Blockson sent a document to the AEC that 

included a best estimate of manufacturing costs, with a breakdown of labor by category. 

Blockson projected the following personnel needs for the uranium operations: 2 Operators per 

shift, 1 Chemist per shift, 2 Daymen, 2 Mechanics (on average), 1 Clerk, 1 Development 

Chemist, and 1 Foreman. From the rate information given and total estimate it could be 

determined that Blockson assumed 4.2 shifts per day, which would allow for 24 hours operations 

every day of the week. This results in about 18 full time personnel and two part time personnel 

according to pre-operational estimates (Blockson 1951). 

In September 1953, after start up of operations, the AEC received a request for bioassay analysis 

services for “about twenty production workers engaged in uranium processing at Blockson 

Chemical Company” (AEC 1953). The first known sample results from these services were 

reported in April 1954 by the AEC New York Operations Office Health and Safety Laboratory 
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(HASL). Subsequently, bioassay results on nine other occasions were reported by HASL 

through February 1958. Sample results are available for twenty five different workers. 

The urinalysis records were found on reports from the AEC New York Operations Office, Health 

and Safety Division. One hundred twenty two sample results are available with the results 

ranging from 0 to 17 μg uranium per liter. The analyses were performed by fluorimetry. Of the 

twenty five workers, nineteen workers had multiple bioassay results, with six workers having a 

single sample result reported. The nineteen workers with multiple results were used to determine 

a distribution of inhalation intake rates. One of the nineteen workers had multiple samples over 

two different time periods with a two year break without urinalysis. For purposes of estimating 

worker intakes, that workers results were analyzed as two workers, resulting in a distribution of 

intake rates for twenty workers. The workers with only one sample result were not used, 

however, those six results ranged from 0 to 6 μg per liter, which fit in the distribution of bioassay 

results. 

Some of the names on the urinalysis reports have been matched with names of some workers 

whose jobs in Building 55 during uranium recovery work are known. These workers include two 

process operators, two chemists/analysts, and a supervisor. Some of these workers have been 

interviewed to confirm the work they performed in Building 55. Their bioassay results support 

the exposure assumptions in this evaluation, i.e., the highest exposed of those known workers 

was an operator whose job included drumming dried yellowcake, and the default intake 

recommendations are favorable in comparison to the data for those workers. 

The International Commission on Radiological Protection (ICRP) recommends three material 

types for solubility of inhaled uranium, Types F, M and S, based on the clearance rate from the 

lungs (ICRP 1994a). Various studies have shown that U3O8 closely corresponds to the clearance 

rate associated with material Type M. Some studies have also shown that high fired material can 

produce uranium compounds that clear more slowly from the lungs, i.e., indicative of material 

Type S (Rucker, et al. 2001). Type M uranium is the most appropriate lung solubility material 

type based on the process used at Blockson. The U3O8 product was produced from wet 

phosphoric acid by filtering the precipitated uranium and then using a dryer to dewater the solids 

(Blockson 1953). No calcining or high firing of the uranium material was indicated by the 

process used at Blockson. Based on these processes and the results of various studies that have 

been summarized by Rucker, et al., Type M material is used to derive intakes from bioassay 

results. 

The Blockson bioassay results were reported in μg per liter, and were converted to μg per day by 

multiplying by a daily excretion rate of 1.4 liters, then converted to pCi/day by multiplying by 

0.677 pCi per μg of natural uranium. Individual worker intakes were determined using IMBAExpertTM by assuming a chronic inhalation intake of Type M uranium with parameters 

recommended by the ICRP (ICRP 1994b). 

For daily intake rate calculation purposes, intakes were assumed to have occurred beginning in 

the year sampled and ending with the last sample date. The results were given an absolute error 

of 1.0, which equally weights the sample results for fitting purposes in IMBA-Expert. Daily 

intake rates ranged from 6 to 76 pCi/day. The intake rate results fit well to a lognormal 
 

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distribution having a median value of 25 pCi/day with a geometric standard deviation of 2.1, as 

shown in Figure 3. The analysis of bioassay results with the assumption of 100% of the intake 

via inhalation is a claimant-favorable assessment of intakes from all pathways except when 

calculating dose to certain tissues of the gastrointestinal tract. Dose from ingestion of uranium is 

discussed below. 

Figure 3: Distribution of uranium intake rates. 

5.00 

4.00 

3.00 

2.00 

1.00 

0.00 

y = 0.7256x + 3.2181 

R2

 = 0.9639 

-2.50 -2.00 -1.50 -1.00 -0.50 0.00 0.50 1.00 1.50 2.00 2.50 

z score 

ln(intake) 

Production workers are assumed to be continually exposed at the 95th percentile intake rate of 82 

pCi/day, and administrative personnel are assumed to be exposed continually at the median 

intake rate of 25 pCi/day. These bounding intakes should be entered into IREP as constants. 

The presence of associated radionuclides that could be present and contribute to significant 

internal dose is assumed (ORAU 2006). The results are summarized in Table 1a below. 

Table 1a: Inhalation rate for operations. 

Worker category Intake rate for Type M material1

 Distribution 

Administrative 25 pCi/day total U Constant value 

Administrative 0.35 pCi/day Th-2282 Constant value 

Administrative 0.35 pCi/day Th-2322

 Constant value 

Production workers 82 pCi/day M total U Constant value 

Production workers 1.1 pCi/day Th-2282

 Constant value 

Production workers 1.1 pCi/day M Th-2322

 Constant value 

1. Intake rates are normalized to units of calendar days. The intake period for operations is March 1, 1951, 

through March 31, 1962. 

2. Thorium intake rates are derived from ratios in ORAU 2006. Solubility types for thorium are based on 

recommendations in IRCP Report 68 (ICRP 1994b). 

Workers also had the potential to ingest uranium from contact with contaminated surfaces or 

from eating or drinking in the area. When deriving intakes from the bioassay results, a chronic 

ingestion of uranium resulted in a higher dose to certain tissues of the gastrointestinal tract when 

compared to the dose from the inhalation intakes described above. Therefore, intakes are 

presented in Table 1b based on the presumption that all the uranium in the workers urine was due 

to ingestion. Although inhalation is the most common mode of intake in a production facility the 
 

 

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presumption of the ingestion pathway provides an upper bounding value for dose from ingestion. 

A worker should be assigned intakes from inhalation or ingestion, not both. 

Table 1b: Ingestion rate for operations. 

Worker category Intake rate Type M Material1,3 Distribution 

Administrative 83 pCi/day total U Constant value 

Administrative 1.2 pCi/day Th-2282 Constant value 

Administrative 1.2 pCi/day Th-2322

 Constant value 

Production workers 270 pCi/day total U Constant value 

Production workers 3.6 pCi/day Th-2282 Constant value 

Production workers 3.6 pCi/day Th-2322 Constant value 

1. Intake rates are normalized to units of calendar days. The intake period for operations is March 1, 1951, 

through March 31, 1962. 

2. Thorium intake rates are derived from ratios in ORAU 2006. 

3. Ingestion intakes provide bounding dose to the stomach, small intestine, upper large intestine, lower large 

intestine, and colon. The f1 values are 0.02 for uranium ingestions and 0.0005 for thorium ingestions. 

3.2 Radon Exposures 

Radon exposures to workers from uranium extraction work at phosphate plants have been 

evaluated for the NIOSH EEOCIPA dose reconstruction project (ORAUT 2006). For 

reconstructing lung doses, all workers at Blockson are to be assigned an exposure of 0.036 WLM 

(working level month) per year due to radon progeny. 

Table 2: Radon exposures. 

Dose component Annual dose/exposure1 Distribution 

Radon progeny 0.036 WLM (lungs only) Lognormal, GSD=2.0 

Radon progeny 75 rem alpha (ET1 only)2 Lognormal, GSD=2.0 

Radon progeny 0.30 rem alpha (ET2 only)2 Lognormal, GSD=2.0 

Radon gas 0.002 rem alpha (nonrespiratory tract tissues only) Constant value 

1. Exposure and dose values from ORAU 2006. 

2. ET1 and ET2 dose conversion factors from NIOSH 2006. 

4.0 External Dose 

External dosimetry data is not known to exist for Blockson workers, and data capture efforts for 

the EEOICPA dose reconstruction project have not found any direct radiation survey results 

from the Blockson facility. Therefore, source term information has been used to estimate 

external doses to workers. Blockson’s uranium recovery process was a by-product process 

designed to fit into the existing phosphate process (Stolz, Jr. 1958). The primary radionuclides 

of interest for potential external exposure in Building 55 are U-238 and daughter radionuclides 

Th-234 and Pa-234m. 

At the Blockson facility, a side-stream of the phosphoric acid was diverted to Building 55 where 

the uranium was separated (Wimpfen 2002). This phosphoric acid was an intermediate product 

in Blockson’s normal commercial production of technical phosphates (DOE 1983). In the 
 

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manufacture of this phosphoric acid, phosphate rock was digested with sulfuric acid, resulting in 

phosphoric acid and phosphogypsum. The uranium remains with the phosphoric acid and the 

radium preferentially follows the phosphogypsum (Roessler et al. 1979, Laiche and Scott 1991). 

For the purpose of dose reconstruction, it is assumed that there was a potential for external 

exposure from the following sources: 

- Submersion in air contaminated with yellowcake dust, 

- Barrels of yellowcake, 

- Contaminated surfaces, 

- Medical x-rays. 

Based on an assumed limiting air concentration of 8.5 pCi/m3

 derived from the 95th percentile 

daily intake rates discussed above, the external dose from submersion in air contaminated with 

uranium dust was calculated based on dose coefficients for U-238 and daughter radionuclides 

Th-234 and Pa-234m from Federal Guidance Report No. 12 (EPA 1993). All organ doses are 

less than 0.001 rem per year. These are insignificant in comparison to the favorable evaluation 

of other dose components and, therefore, are not included in dose reconstructions. 

Dose from occupationally required medical X-rays have also been considered and assumed to 

have occurred, although no information has been found to indicate that Blockson or the AEC 

required X-rays of the workers. 

4.1 Source Term 

Clegg and Foley (1958) state that freshly separated yellowcake has a very low gamma emission 

rate; therefore external radiation is of no particular concern at this stage of the process. 

However, due to ingrowth of daughter radionuclides in the yellowcake, the radiation levels 

increase for several months following production (NRC 2002b). 

For accumulations of processed yellowcake dust, the surface beta dose rate from U-238 

daughters is negligible just after separation, but rises steadily until Pa-234m and Th-234 reach 

equilibrium concentrations. After a few months, the beta surface dose rate is about 150 mrem/hr 

(NRC 2002a). Figure 4 shows the rise in beta dose rate during 100 days after separation from 

ore. 

Figure 5 shows that the beta dose rate from the surface of yellowcake decreases rapidly as a 

function of distance from the surface. Rapid decrease in the beta dose rate with distance, and the 

shielding afforded by shoes and clothing, reduces dose from beta radiation, particularly from 

yellowcake deposited on floors. 
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Figure 4. Beta dose rate on the surface of yellowcake. [Reproduced from 

NRC 2002a] 

Figure 5. Beta dose rate from yellowcake separated from ore for more than 

100 days as a function of distance from the surface. [Reproduced 

from US NRC 2002a] 

4.2 Exposure from Drums of Uranium 

MCNPX (version 2.5.0) was used to determine the dose rate per curie of 238U regardless of the 

actual activity in the drum. This was later adjusted for actual source activity to compare actual 
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dose rates. All radionuclides were ratioed with respect to 238U to determine the number of 

photons and electrons per decay of 238U. Anderson and Hertel (Anderson and Hertel 2005) 

showed that the short lived nuclides (234Th, 234mPa, 234Pa, and 231Th) are very close to equilibrium 

(adjusted for branching ratios) at 100 days. For the purposes of this evaluation, branching ratio 

adjusted equilibrium was assumed. ICRP Publication 74 Table A.1 was used to convert the 

photon flux to units of air kerma using the conversion factors in Table 3 below (ICRP 1996; 

Stabin and da Luz 2002). 

Table 3: Activity of aged uranium following separation and 100 day ingrowth. 

 Activity Concentration 

in aged Uranium 

Metal (Bq/g) 

Relative activity 

concentration 

(normalized to 238U)

238U 12200 1 

234Th 12200 1 

234mPa 12200 1 

234Pa 19.52 0.0016 

234U 12200 1 

235U 555 0.045492 

231Th 555 0.045492 

Dose 30 cm from drum 

The dose rate was determined at 77.9 cm above the ground, 30 cm from the edge of the drum for 

both the photon and beta emissions of natural uranium and its progeny. Results of these are 

provided in Table 4 and Figure 6. 

Table 4: Uranium dose rates from drums of yellowcake. 

Density of 

U3O8 (g cm-3) 

Activity of U in 

drum (Ci) 

Photon 

emission 

dose(rad/hr) 

Bremsstrahlung 

dose (rad/hr) 

Total dose rate 

at 30 cm 

(rad/hr) 

0.5* 3.121E-02 3.96E-04 3.20E-4 7.16E-4 

1 6.242E-02 5.00E-04 3.60E-04 8.60E-04 

2 1.248E-01 5.54E-04 3.76E-04 9.30E-04 

4 2.497E-01 5.84E-04 3.84E-04 9.69E-04 

6 3.745E-01 5.84E-04 3.64E-04 9.48E-04 

6.7 4.182E-01 5.81E-04 3.74E-4 9.56E-4 

*The drum begins to noticeably impact the dose rates at low material concentration. 

The affect of density of the drummed uranium concentrate on the modeled dose rate was 

evaluated. The effective density of the drummed material was assumed to be variable up to 

6.7 g cm-3 . The results are shown in Figure 6. 

Based on these results, the effective density of drummed U3O8 concentrate makes little 

difference in the calculated dose rates from 1 to 6.7 g cm-3 . This would not be true if significant 

contaminants were introduced. Assuming the material is pure U3O8 with a density of 4 g cm-3 is 

favorable to claimants without excessive bias because of the self shielding created by additional 

materials. The introduction of significant quantities of additional materials essentially will 
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provide additional shielding and reduce the effective Z of the material, both of which will reduce 

the dose rate. Furthermore, as is evidenced by the drop at 0.5 g cm-3, the steel shell of the drum 

makes a bigger impact on the dose results for low density uranium. This dose rate compares well 

with the result obtained by Anderson et al. (2005) from a bare, cylindrical uranium ingot at 30.48 

cm (16.51 cm radius, 50.8 cm height) for a total of 1.15 mrem/hour. Differences arise from 

changes to geometry and shielding from the drum walls. 

Figure 6: Effect of density on dose rate at 30 cm from drum of U3O8. rad hr-1 

0.0000 

0.0002 

0.0004 

0.0006 

0.0008 

0.0010 

0.0012 photons 

bremsstrahlung 

total 

0 1 2 3 4 5 6 7 8 

density (g cm-3) 

Assignment of a 50%/50% distribution of dose from 30-250 keV photons and >250 keV photons 

was verified to be favorable to claimants as only approximately 8% of the total dose comes from 

photons less than 300 keV (originating from both bremsstrahlung and photon emissions) in a 

drum with density of 4 g/cm3 . Cumulative energy distribution results are shown in Figure 7. 
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Figure 7: Cumulative dose rate from photons and bremsstrahlung at 30 cm from 

a drum containing U3O8 (density 4 g/cm3). 

0.0012
 

0.0010
 

0.0008
 

cumulative rad/hr 0.0006 

0.0004 

0.0002 

0.0000 

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 

Photon Energy (MeV) 

The air kerma dose rates were converted to annual organ doses by assuming a worker’s exposure 

time was lognormally distributed. The median exposure time was determined by assuming all 

workers were working eight hours per day, one day per week at a distance of 1 foot from the 

drum. This was normalized to 400 hours per work year. The 95th percentile exposure time was 

determined by assuming the worker spent a standard 2000 hour work year at a distance of 1 foot 

from the drum. This results in a kerma dose distribution with a median value of 0.387 rad per 

year with a geometric standard deviation of 2.7. The dose distribution was assumed to be 50% 

from photons 30-250 keV and 50 % from photons > 250 keV. 

For the purposes of calculating organ doses for use in the NIOSH Interactive 

RadioEpidemiological Program (NIOSH-IREP) to the whole body doses were input into 

CRYSTAL BALL® using Monte Carlo methods to multiply times the triangular organ dose 

conversion factors for kerma to organ dose found in NIOSH External Dose Reconstruction 

Implementation Guideline (NIOSH 2002). The results are annual doses that are lognormally 

distributed. The results are in Table 5. For skin, the air kerma values were multiplied by 1.0. 
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Table 5: Annual organ dose from exposure to drums of yellowcake. 

30-250 keV photons >250 keV photons 

Organ dose, rem GSD dose, rem GSD 

Bladder 0.223 2.8 0.213 2.7 

RBM 0.100 2.9 0.163 2.7 

Bone 

Surface 0.266 2.8 0.179 2.8 

Breast 

(female) 0.264 2.6 0.223 2.7 

Colon 0.179 2.8 0.198 2.7 

Esophagus 0.107 2.9 0.167 2.7 

Eye 0.265 2.7 0.215 2.7 

Ovaries 0.158 2.8 0.193 2.7 

Testes 0.300 2.7 0.225 2.7 

Liver 0.123 2.7 0.166 2.7 

Lung 0.171 2.8 0.200 2.6 

Remainder 0.146 2.8 0.185 2.6 

Stomach 0.232 2.7 0.213 2.7 

Thymus 0.263 2.7 0.224 2.7 

Thyroid 0.285 2.7 0.228 2.7 

Uterus 0.174 2.8 0.188 2.7 

Skin 0.194 2.7 0.194 2.7 

4.3 Exposure from Contaminated Surfaces 

Estimates of external dose from surfaces contaminated with uranium have been performed. The 

95th percentile intake rates from inhalation were used to derive an average airborne 

concentration of 8.54 pCi/m3 . A terminal settling velocity of 0.00075 m/s was used as an 

estimate of the velocity of deposition to surfaces in the building. The value is within the range of 

deposition velocities measured in various studies (NRC 2002b). It was assumed that uranium 

settled on plant surfaces at a steady state 24 hours per day for 365 consecutive days with no 

cleanup or removal of contamination. 

The estimated surface contamination results for U-238 and associated daughter radionuclides 

Th-234 and Pa-234m were multiplied by the Dose Coefficients for Exposure to Contaminated 

Ground Surface found in Table III.3 of Federal Guidance Report No. 12 (EPA 1993). With the 

exception of dose to the skin, the annual dose for all organ is less than 0.001 rem and is 

insignificant in comparison to the favorable assignment of dose from drums of yellowcake. Skin 

dose is discussed in section 4.4 below. 

Although the modeled external doses are assumed to be insignificant based on airborne uranium 

concentrations there could have been localized spots with greater potential for exposure. The 

only direct reading radiation results available are those from surveys done from March to 

November 1978 by Argonne National Laboratory (ANL) (DOE 1983). 
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The ANL survey was performed throughout Building 55, including plant surfaces, tanks, pipes, 

and other process equipment. The report estimated that some areas were not accessible, but that 

they had surveyed an estimated 95% of the floors and 90% of the walls. The surveys included 

contamination surveys. A dose rate was taken at contact and at 1 meter on all 63 spots in which 

contamination was detected. The dose rates at 1 meter on 7 of the 63 “hot” spots ranged from 

0.04 mR/hr to 0.2 mR/hour. The other 56 spots had 1 meter dose rates indistinguishable from 

background. The reported background dose rate on the instrument used was between 0.02 mR/hr 

and 0.03 mR/hr. The results of the 7 spots with measurable 1 meter dose rates included the 

background dose rates. From a review of the survey map and results it seems improbable that a 

worker could be significantly exposed above the background rate of 0.03 mR/hr for significant 

time. However, in the absence of individual dosimeter data, whole body dose rates are modeled 

by a lognormal distribution with the 0.03 mR/hr rate assumed to be the median value for 2,000 

hours per year. The geometric standard deviation is 3.2, which was determined by assuming that 

the 95th percentile dose rate is equal to the maximum observed result of 0.2 mR/hr. 

There are no survey results to use to assign external doses until the 1978 survey by ANL. The 

1978 survey suggests that either the facility was only contaminated in localized spots, or it 

indicates that some areas of surface contamination was not subject to the same contamination 

depletion rate that would be expected from natural processes and from general cleaning and 

weathering that would have occurred since 1962. The 7 spots were on small areas of floor 

(reported to be about 0.5 m2

 each) and on a pipe inlet and on a spot on a machine. Additionally, 

since the normal non-AEC related operations at Blockson produced byproduct that contained 

small amounts of uranium and daughter products, it is unknown if that would have contributed to 

the contamination of the facility after AEC operations ended in 1962. Based on the above 

considerations and the absence of more data, an assumption that is favorable to claimants is 

made for deriving dose rates prior to 1978. It is assumed that the few spots in the facility with 

measurable dose rates were representative of the entire facility from the operational period to the 

present or until the end of the workers’ EEOICPA covered employment. The deep dose rates 

during the operational period are adequately modeled by the assumptions made in deriving doses 

from drums of yellowcake. Therefore, dose from contaminated surfaces is only assigned during 

the residual contamination period. 

4.4 Beta Dose 

For accumulations of processed yellowcake dust, the surface beta dose rate from U-238 

daughters is negligible just after separation, but rises steadily until Pa-234m and Th-234 reach 

equilibrium concentrations. After a few months, the beta dose rate is about 150 mrem/hr (NRC 

2002a). Figure 4 shows the rise in beta dose rate during 100 days after separation from ore. 

It is assumed that there was a potential to receive a shallow dose from exposure to open drums of 

yellowcake during drum loading and sealing. According to Figure 5 the dose rate at 1 foot from 

the surface of aged yellowcake is between 1 and 2 mrem/hour. It is assumed that the production 

workers spent 8 hours per week, 50 weeks per year, at 1 foot from the surface of aged 

yellowcake at a dose rate of 2 mrem/hour. This results in a shallow beta dose of 0.8 rem/year. 

To allow for uncertainty, the time of exposure was assumed to be lognormally distributed with 

the 95th percentile exposure time assumed to 40 hours per week, 50 weeks per year. This results 
 

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in a shallow dose of 0.8 rem per year. Results for dose reconstructions are in Table 6. The 

calculated beta doses have not been reduced to allow for doses to areas of the skin that are 

typically covered by clothing resulting in a reduction of the beta dose to the skin. 

Table 6: Shallow dose. 

Dose component Annual dose/exposure1

 Distribution 

Beta dose, E>15 keV 0.8 rem per year Lognormal, GSD=2.7 

4.5 Occupational Medical Dose 

Dose from occupationally required medical X-rays have also been considered and assumed to 

have occurred, although no information has been found to indicate that Blockson or the AEC 

required X-rays of the workers. For the AEC operational period at Blockson, employees are 

assumed to have received an annual chest X-ray. Organ doses are listed in Table 7 below and 

are based on an assumed Posterior-Anterior (PA) exposure with minimal collimation. Dose 

values are reproduced from Table 6-5 of “Dose Reconstruction from Occupationally Related 

Diagnostic X-ray Procedures” (ORAU 2005.). The annual doses are applied as dose from 30­

250 keV photons using the values in Table 7 as the mean of a normal distribution with a 30% 

standard deviation. 

Table 7: Annual organ dose from medical X-rays. 

Organ 

Annual dose, rem 

photon 30-250 keV 

Thyroid 3.48E-02 

Eye/brain 6.40E-03 

Ovaries 2.5E-02 

Liver/gall bladder/spleen 9.02E-02 

Urinary bladder 2.5E-02 

Colon/rectum 2.5E-02 

Testes 5.0E-03 

Lungs (male) 8.38E-02 

Lungs (female) 9.02E-02 

Thymus 9.02E-02 

Esophagus 9.02E-02 

Stomach 9.02E-02 

Bone surfaces 9.02E-02 

Remainder 9.02E-02 

Breast 9.80E-03 

Uterus 2.5E-02 

Bone marrow (male) 1.84E-02 

Bone marrow (female) 1.72E-02 

Skin 2.70E-011 

1. Skin dose is for skin in the primary beam. 
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5.0 Dose from Residual Contamination 

The whole body median dose rate of 0.060 R/year derived from the discussion in section 4.3 

above was used to calculate various organ doses. Photon dose is divided equally between 30-250 

keV and the >250 keV ranges. The Roentgen to organ dose conversion factors for isotropic 

geometry in the External Dose Reconstruction Implementation Guideline (NIOSH 2002) were 

used to estimate organ doses. Skin doses were calculated assuming a conversion factor of 1.00. 

Results are in Table 8. 

Table 8: Annual dose from residual contamination.1 

Photons 

E=30-250 keV 

Photons 

E>30-250 keV 

Organ dose, rem dose, rem 

Bladder 1.61E-02 1.94E-02 

RBM 1.67E-02 2.00E-02 

Bone Surface 2.81E-02 2.04E-02 

Breast 2.12E-02 2.22E-02 

Colon 1.55E-02 1.90E-02 

Esophagus 1.50E-02 1.96E-02 

Eye 2.23E-02 2.28E-02 

Ovaries 1.48E-02 1.88E-02 

Testes 1.90E-02 2.08E-02 

Liver 1.70E-02 2.00E-02 

Lung 1.88E-02 2.12E-02 

Remainder 1.67E-02 1.99E-02 

Stomach 1.70E-02 1.99E-02 

Thymus 1.84E-02 2.05E-02 

Thyroid 1.92E-02 2.14E-02 

Uterus 1.46E-02 1.81E-02 

Skin 3.00-02 3.00E-02 

1. For dose reconstructions, annual doses are applied as 

lognormal distributions with a GSD of 3.2. 

Uranium inhalation intakes during the residual contamination period have been derived from the 

operational period intakes and from estimated airborne radioactivity derived from the 1978 

FUSRAP survey. While the uranium recovery operations could result in high localized air 

concentrations, air concentrations from resuspension of residual contamination would be more 

consistent throughout the area. Interviews with former workers indicate that housekeeping was 

performed regularly to reduce build up of material on the floors. After cessation of uranium 

recovery work the main source of contamination (precipitated and dried yellowcake) was no 

longer present. Therefore, the derived median inhalation rate of 25 pCi/day is used as the 

inhalation intake rate at the start of the residual contamination period on April 1, 1962. 

Thereafter, airborne radioactivity from resuspension of contamination in the facility and 

corresponding intakes are assumed to decrease according to an exponential model described 

below. 
 

 

 

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Extensive radiological surveys were performed starting in March 1978. Airborne radioactivity 

concentration at 16 years post operation was estimated by assuming that the facility was 

uniformly contaminated at the level of the maximum alpha smear result of 640 dpm/100 cm2

 as 

reported in the 1978 survey (DOE 1983). This value was multiplied by a resuspension factor of 

1E-06 m-1 (NRC 2002c), which results in an estimated maximum residual air concentration of 

0.03 pCi/m3 . Applying a breathing rate of 1.2 m3/hr for 2000 hours per year results in a potential 

inhalation of about 0.2 pCi/day at the time of the March 1978 survey. The 1962 and 1978 daily 

intake rates were used to estimate the rate of reduction of intakes according to the following 

equation. 

It = I 0 * e−λ t 

where: 

It = daily intake rate at time t 

t = time (days) since April 1, 1962 

I0 = daily intake on April 1, 1962 

e = base of the natural logarithms 

λ = exponential constant 

The derived intake of 25 pCi/day on April 1, 1962, was substituted for I0. The 0.2 pCi/day 

derived intake from the 1978 survey was substituted for It on April 1, 1962. The time between 

April 1, 1952, and April 1, 1962 is 5844 days. This resulted in the following equation to 

calculate the exponential constant λ. 

0.2 pCi 25 pCi −λ*5844 d = * e

d d 

The constant λ was determined to be 0.000826 day-1 . 

Average daily inhalation intake rates for each year between 1962 and 1978 were then calculated 

and are given in Table 9a below. The 1978 intakes are relatively low and are applied for all 

subsequent years. The methods used for derivations of these intakes are considered bounding, 

and the corresponding annual doses are considered constants for purposes of dose reconstruction. 
 

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Table 9a: Inhalation intake rate from residual contamination1 

Year 

Uranium Type M 

pCi/day 

Th-228 Type M 

pCi/day 

Th-232 Type M 

pCi/day 

1962 

(Apr. 1 – Dec. 31) 22 0.31 0.31 

1963 17 0.24 0.24 

1964 13 0.17 0.17 

1965 9.4 0.13 0.13 

1966 7.0 0.10 0.10 

1967 5.1 0.070 0.070 

1968 3.8 0.052 0.052 

1969 2.8 0.039 0.039 

1970 2.1 0.028 0.028 

1971 1.5 0.021 0.021 

1972 1.1 0.016 0.016 

1973 0.84 0.012 0.012 

1974 0.62 0.009 0.009 

1975 0.46 0.006 0.006 

1976 0.34 0.005 0.005 

1977 0.25 0.003 0.003 

1978 – end 0.19 0.003 0.003 

1. Inhalation intakes are not assigned for calculating dose to the stomach, small intestine, 

upper large intestine, and lower large intestine. See Table 9b for ingestion intakes for 

those tissues. 

Inhalation of uranium is considered to be proportional to the airborne concentration. For the 

residual period this is a function of the amount of loose contamination present on plant surfaces. 

Likewise, the potential for ingestion of uranium is a function of the amount of loose 

contamination present on plant surfaces. This indicates a relationship between airborne 

concentration and ingestion of material. Bounding ingestion intakes were evaluated and 

discussed above for the AEC operational period. Based on the median intake rate, that 

evaluation indicates that doses to the gastrointestinal tract are bounded by an ingestion of 

83 pCi/day, which is based on the evaluation of bioassay results under the assumption that no 

inhalation occurred. 

To estimate bounding intakes for ingestion during the residual contamination period the daily 

ingestion rate is reduced at the same rate as the inhalation intakes described above. This results 

in the following expression to determine ingestion intakes during the residual contamination 

period. 
 

 

 

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83 pCi −0.000826d *t It = * e −1 

d 

where: 

It = daily ingestion intake rate at time t 

 83 pCi/d = bounding ingestion intake on April 1, 1962 

t = days since April 1, 1962 

The above equation was used to derive average daily ingestion intake rates for each year between 

1962 and 1978. Results are presented in Table 9b. The ingestion intake rate is based on 

bounding dose to the stomach, small intestine, upper large intestine, and lower large intestine. 

The modeled ingestion intake rate assumes that all material assimilated from contaminated 

surfaces is via the ingestion pathway. Therefore, inhalation intakes are not assigned for these 

tissues. The doses are considered constants for dose reconstruction purposes. 

Table 9b: Ingestion intake rate from residual contamination1 

Year 

Uranium Type M 

pCi/day 

Th-228 Type M 

pCi/day 

Th-232 Type M 

pCi/day 

1962 

(Apr. 1 – Dec. 31) 74 1.0 1.0 

1963 57 0.78 0.78 

1964 42 0.58 0.58 

1965 31 0.43 0.43 

1966 23 0.32 0.32 

1967 17 0.23 0.23 

1968 13 0.17 0.17 

1969 9.3 0.13 0.13 

1970 6.9 0.095 0.095 

1971 5.1 0.070 0.070 

1972 3.8 0.052 0.052 

1973 2.8 0.038 0.038 

1974 2.1 0.028 0.028 

1975 1.5 0.021 0.021 

1976 1.1 0.016 0.016 

1977 0.83 0.011 0.011 

1978 – end 0.62 0.008 0.008 

1. Ingestion intakes are assigned only when calculating dose to the stomach, small 

intestine, upper large intestine, and lower large intestine. No inhalation intake is 

assigned for these tissues. The f1 values are 0.02 for uranium ingestions and 0.0005 

for thorium ingestions. 
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6.0 Dose Reconstruction Summary 

For EEOICPA dose reconstruction purposes, exposure starts March 1, 1951, or the first date the 

employee has covered employment at Blockson, whichever is later. The end of the operational 

period is March 31, 1962. Residual contamination doses start on April 1, 1962. 

External doses are assigned according to the values specified in Tables 5, 6, 7, and 8, as 

applicable to the tissue of concern. The external dose values in Tables 5 and 6 are normalized 

annual doses and need to be fractioned for partial years worked. For external dose from residual 

contamination, the values in Table 8 are also normalized annual doses and need to be fractioned 

for partial years worked. The occupational medical x-ray doses in Table 7 are to be assigned by 

assuming one exposure per year of covered employment during the operational period. 

In the absence of worker-specific bioassay data, the default intakes are assigned. For dose to the 

stomach, small intestine, upper large intestine, and lower large intestine, intakes are assigned 

from Tables 1b, 2, and 9b. For all other tissues assign intakes from Tables 1a, 2, and 9a. Natural 

uranium doses can be calculated by assuming 50% of the intake is from U-234 and 50% is from 

U-238. 
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7.0 References 

AEC (U. S. Atomic Energy Commission) 1951, Contract for Uranium Production, Contract No. 

AT- (49-1)-611, October 18, 1951 [SRBD Ref ID 16328]. 

AEC 1953, letter from James A. Barr, Jr., Assistant Director, Domestic Production, Division of 

Raw Materials, to Dr. Richard Fahrner, Medical Consultant, Blockson Chemical Company, 

September 14, 1953 [SRBD Ref ID 9728, p. 16]. 

AEC 1955a, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates, Office of Domestic Procurement,” Washington D.C., August 31, 1955 [SRDB Ref 

ID 22264]. 

AEC 1955b, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates, Office of Domestic Procurement,” Washington D.C., September 30, 1955 [SRDB 

Ref ID 22264]. 

AEC 1955c , Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates, Office of Domestic Procurements,” Washington D.C., October 31, 1955 [SRDB Ref 

ID 22264]. 

AEC 1955d, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates, Office of Domestic Procurement,” Washington D.C., November 30, 1955 [SRDB 

Ref ID 22264]. 

AEC 1955e, Division of Raw Materials, “Monthly Report on Activities of Domestic Production 

Phosphates,” Office of Domestic Procurement, Washington D.C., December 1955 [SRDB Ref 

ID 22264]. 

Anderson, J.L., and Hertel, N.E., “Bremsstrahlung Doses from Natural Uranium Ingots,” 

Radiation Protection Dosimetry, 115 (1-4): 298-301, 2005. 

Barr, J.A., Jr., Ruch, J.W., and Borlik, R.F., “Recovering Uranium As By-Product in Phosphate 

Processing,” presented before the annual meeting of the American Chemical Society, Chicago, 

and reprinted from Rock Products, October 1955. 

Blockson (Blockson Chemical Company) 1951, letter from E. B. Lopker, Director of 

Engineering, to Sheldon Wenpfen, Atomic Energy Commission, Raw Materials Operations, July 

31, 1951 [SRDB Ref ID 9558]. 

Blockson 1953, “Uranium Recovery from Wet Process Phosphoric Acid, Abstract,” prepared by 

Stolz, Emil Jr., February 23, 1953 [SRDB Ref ID 4027]. 

Clegg, J.W., and Foley, D.D., “Uranium Ore Processing,” prepared under contract with the U.S. 

AEC, Addison-Wesley Publishing Company, Inc., 1958, pp. 375-378. 

DOE (U. S. Department of Energy) 1983, FUSRAP, “Radiological Survey of Chemicals Group, 

Olin Corporation (formerly Blockson Chemical Company, Joliet, Illinois, March 27 – November 
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28, 1978,” Argonne National Laboratory; DOE/EV-0005/35; ANL-OHS/HP-83-103, May 1983 

[SRDB Ref ID 23615]. 

DOE 1985, FUSRAP Authority Review of the Former Blockson Chemical Company, Joliet, 

Illinois, 1985 [SRDB Ref ID 11587]. 

DOE 2001, DOE Standard: Health Physics Manual of Good Practices for Uranium Facilities, 

(DOE-STD-1136-2000, Change Notice No. 3, December 2001). 

http://tis.eh.doe.gov/techstds/standard/std1136/STD11362000.pdf 

DOE 2006, Worker Advocacy Web Site Facility List, June 8, 2006, 

http://www.eh.doe.gov/advocacy/faclist/findfacility.cfm 

Eidson, A.F., and Damon, E.G., “Predicted Deposition Rates of Uranium Yellowcake Aerosols 

Sampled in Uranium Mills,” Health Physics, 46(1): 165-176, 1984. 

EPA (U. S. Environmental Protection Agency) 1993, “External Exposure to Radionuclides in 

Air, Water, and Soil,” Federal Guidance Report No. 12, September 1993. 

ICRP (International Commission on Radiological Protection) 1994a, “Human Respiratory Tract 

Model for Radiological Protection,” ICRP Publication 66. 

ICRP 1994b, “Dose Coefficients for Intakes of Radionuclides by Workers,” ICRP Publication 

68. 

ICRP 1996, “Conversion Coefficients for Use in Radiological Protection Against External 

Radiation, ICRP Publication 74. 

Kocher, D.C., and Eckerman, K.F., “Electron Dose-Rate Conversion Factors for External 

Exposure of the Skin from Uniformly Deposited Activity on the Body Surface,” Health Physics, 

53(2): 135-141. 

Laiche, T. P., and Scott, L. M., “A Radiological Evaluation of Phosphogypsum,” Health Physics, 

Vol. 60(5):691-693, 1991. 

McGinley, Frank, Personal Communication, December 2002. 

NIOSH (National Institute for Occupational Safety and Health) 2002, External Dose 

Reconstruction Implementation Guideline, OCAS-IG-001, Rev. 1, Office of Compensation 

Analysis and Support, Cincinnati, Ohio. 

NIOSH 2006, “Lung Dose Conversion Factor for Thoron WLM,” OCAS-TIB-0011, Rev. 2, 

Office of Compensation Analysis and Support, Cincinnati, Ohio, January 13, 2006. 

NRC (U. S. Nuclear Regulatory Commission) 2002a, “Health Physics Surveys in Uranium 

Recovery Facilities,” Regulatory Guide 8.30, Revision 1, May 2002. 

NRC 2002b, “Technical Basis for Calculating Radiation Doses for the Building Occupancy 

Scenario Using the Probabilistic RESRAD-Build 3.0 Code, NUREG/CR-6755, February 2002. 
Effective Date: 

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NRC 2002c, “Re-evaluation of the Indoor Resuspension Factor for the Screening Analysis of the 

Building Occupancy Scenario for NRC’s License Termination Rule, NREUG-1720, Draft Report 

for Comment, June 2002. 

ORAU (Oak Ridge Associated University Team) 2005, “Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures,” ORAUT-OTIB-0006, Rev 03 PC-1, 

December 21, 2005. 

ORAU 2006, “Characterization of Occupational Exposures to Radium and Radon Progeny 

During Recovery of Uranium from Phosphate Materials, ORAUT-OTIB-0043, Rev 00, January 

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